IR 05000271/1987004
| ML20206L646 | |
| Person / Time | |
|---|---|
| Site: | Vermont Yankee File:NorthStar Vermont Yankee icon.png |
| Issue date: | 04/09/1987 |
| From: | Elsasser T NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20206L586 | List: |
| References | |
| 50-271-87-04, 50-271-87-4, IEIN-85-045, IEIN-85-45, NUDOCS 8704170191 | |
| Download: ML20206L646 (16) | |
Text
o U. S. NUCLEAR REGULATORY COMMISSION
REGION I
Report No.
87-04 Docket No.
50-271 License No.
DPR-28 Licensee:
Vermont Yankee Nuclear Power Corporation RD 5, Box 169, Ferry Road Brattleboro, Vermont 05301 Facility:
Vermont Yankee Nuclear Power Station Location:
Vernon, Vermont Dates:
February 3 - March 2, 1987 Inspector:
William J. Raymon,
ior Resident Inspector Approved by:
/W O
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Thomas C. Elsasse M bief, Reactor Projects Section 3C date Inspection Summary: Inspection on February 3 - March 2,1987 (Report No.
50-271/87-04)
Areas Inspected:
Routine, unannounced inspection on day time and backshifts by the resident inspector of:
actions on previous inspection findings; physical security; plant operations; actions in response to the discovery of contaminated material in a sanitary dumpster; maintenance activities; audit program results; staffing changes; and, followup of Information Notice 85-45. The inspection in-volved 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
Results: No violations were identified in 7 ef 8 areas inspected.
Routine reviews of plant activities identified no conditions adverse to safe plant operations.
One violation was identified in the maintenance area concerning the failure to control maintenance and modification activities on the toxic gas monitoring system per administrative procedure AP 0025. (Section 11.0). A second, licensee-identi-fied violation (not cited) concerned the inadvertent release of contaminated mate-rial outside the radiation controlled area (Section 7.0).
While the specific con-tamination incident was not significant, it appears as another example of similar problems in the control of low level radioactive materials.
8704170191 870 PDR ADOCK 050 0.71 O
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DETAILS 1.0 Persons Contacted Interviews and discussions were conducted with members of the licensee staff and management during the report period to obtain information pertinent to the areas inspected.
Inspection findings were discussed periodically with the management and supervisory personnel listed below.
Mr. P. Donnelly, Maintenance Superintendent Mr. R. Lopriore, Maintenance Supervisor Mr. J. McCarthy, Emergency Plan Coordinator Mr. R. Pagodin, Technical Services Superintendent Mr. J. Pelletier, Plant Manager Mr. R. Wanczyk, Operations Superintendent
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2.0 Summary of Facility Activities The plant continued routine operations at rated power during the inspection
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period. NRC Region I specialist inspectors completed reviews during the period of February 9-13, 1987 of the routine plant health physics program (Inspection 87-03) and the physical security program (Inspection 87-05).
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3.0 Status of Previous Inspection Findings 3.1 (0 pen) Unresolved Item 87-02-04: Battery Room Temperatures.
Licensee maintenance personnel informed the inspector on February 12, 1987 of the results of the preliminary determinations regarding low battery room
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temperatures. After consultation with YNSD engineering personnel, the licensee confirmed that no operability concerns, due to low room tempera-tures, exist for any batteries at the station.
Specifically, the capacity of the "B" station battery, which is loaded to 96% of its design capacity at 77 degrees F, would equal its load at 67 degrees F.
The "B" battery is located near an exterior wall in the battery room and, based on measurement of each cell, has an average tem-
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perature of 70.65 degrees F, with individual cells in the 60 cell battery operating in the range from 66 to 74 degrees F.
The licensee stated further that an engineering evaluation had been in progress prior to the
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NRC review of this matter to more accurately determine the "B" battery capacity margin based on realistic loading conditions.
The revised cal-culations are expected to show additional margin to the "B" battery capacity. The "A" station battery has 12% margin between its load and capacity, and would have an ambient temperature about 3 degrees F higher than the "B" battery due to its interior location in the battery room,.
Relatively low ambient temperatures have been observed for two other batteries systems: AS2, which provides control power for the
"A" emer-gency diesel generator; and, the battery used to start the emergency lighting generator. The licensee stated that no operability concerns
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exist for either battery due to the spare capacity margins that exist -
50% for the emergency lighting generator, and 68% for AS2, respectively.
Based on the above, the inspector had no further questions regarding near term operability issues.
The present battery surveillance program does not include provisions to adjust specific gravity readings for electrolyte level, or temperature correction factors for outage capacity tests.
The service test uses a block loading scheme instead of a load profile, and is run for four hours instead of the design operating period of 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
The test meets regu-latory requirements since it demonstrates that the battery is capable of delivering the total energy required by the design. The inspector noted that the licensee is in the process of revising battery surveil-lance procedures OP 4210 and 4211 in response to recent INPO findings that will upgrade testing requirements to meet IEEE 450 criteria.
This item will be reviewed further pending receipt and review of the revised test procedures.
This item remains open pending further action by the licensee to estab-lish operational temperatures limits for all batteries of safety /regula-tory concern, and additional monitoring of environmental controls as necessary to assure capacity margins are maintained.
3.2 (Closed) Unresolved Item 87-02-02: Revision of LER 86-12.
The Itcensee submitted Revision 2 for LER 86-12 by. letter VYV 87-029 dated February 13, 1987.
The revised LER correctly identified that Exxon Nebula EP-0 grease was used during valve maintenance, and that its less viscous pro-perties contributed to the limitorque failure for valve V2-43B. This item is closed.
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3.3 (Closed) Unresolved Item 87-02-01: LER Timeliness.
During a meeting with the plant manager on February 20, 1986, the inspector expressed concerns regarding the timeliness of submitting LERs based on seven reports that were not submitted within the 30-day period. During a subsequent meeting with the Technical Support Supervisor, the inspector noted that LER 85-13 was submitted within 30 days of the event date of November 27, 1985.
LER 86-14 was submitted 80 days after the event, but only because the licensee's initial determination was that the event was not reportable.
Further review of the LERs that were submitted one to three days late showed that the 30-day due date fell on a weekend or holiday.
For the five LERs in this category, submittal of the LERs in excess of 30 days was acceptable per the clarifications provided in NUREG 1022, Supplement 1, since they were mailed on the first working day following the end of the 30-day period.
Based on the above, with the exception of LER 86-14, the LERs reported were submitted in accordance with acceptable reporting practices. The adequacy of licensee event reporting will be reviewed further as part of future routine inspections. This item is closed.
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3.4 (0 pen) Unresolved Item 86-18-02: Degraded Block Walls. This item was last reviewed in Inspection 86-22.
During a meeting with the maintenance superintendent on February 20, 1987, the licensee provided the following schedule information regarding plant changes to correct deficiencies in turbine building block walls: (1) control cables for the diesel generator room fans will be rerouted during the outage starting in August, 1987; (2) signal cables for the four main steam line monitors will be rerouted closer to the ventilation room East wall during the 1987 outage to take ddVantage of the protection afforded by the steel I-beam separating the rooms; (3) cables for the reactor recirculation units in the ECCS pump rooms (RRUs 5, 6, 7, and 8) will be rerouted during the 1987 outage; and, (4) the block wall adjacent to the SAC-1A unit will be disassembled from the top to assure its failure at its reduced height could not impact the ventilation system. This work could be started prior to the outage under 10 CFR 50.59.
The inspector acknowledged the above information and had no further com-ments at the present time. This item will be reviewed further during a subsequent inspection pending completion of the licensee's actions.
3.5 (0 pen) Unresolved Item 85-22-02: Program for Updated Vendor Information.
This item concerns the NRC finding that the licensee does not have ar active program to assure vendor and equipment supplier information is current and complete.
Instead, the licensee uses the nuclear utility task action committee (NUTAC) vendor equipment technical information program (VETIP) to meet the intent of the requirements of Generic Letter 83-28, " Generic Implementation of the Salem ATWS Events". This program relies upon equipment failure data at Vermont Yankee or other nuclear plants to obtain updated vendor recommended improvements in products, services and equipment maintenance practices.
The inspector discussed this matter with the maintenance superintendent on February 20, 1987 and expressed his concerns regarding the apparent failure of the licensee's vendor information program to provide timely notification related to the following problems: (1) vendor improvements made since 1976 to limitorque motor operated valves to preclude hydraulic lock-up of the torque switch spring pack (Inspection Item 86-10-03); and, (2) vendor recommendations made in 1979 to increase the Bingham residual heat removal and core spray pump minimum flows from about 5% to about 30% of design (Inspection Item 86-25-02).
This item remains open pending further licensee and NRC review of his actions to meet the requirements of Generic Letter 83-28.
4.0 Observations of Physical Security Selected aspects of plant physical security were reviewed during regular and backshift hours to verify that controls were in accordance with the security plan and approved procedures. This review included the following security measures: guard staffing; vital and protected area barrier integrity; main-
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tenance of isolation zones; and, implementation of access controls, including authorization, badging, escorting, and searches.
No inadequacies were iden-tified, except as discussed below.
The inspecto" reviewed security event report No. 87-01 dated February 11, 1987, which described an event that occurred on February 3, 1987.
The event was previously reviewed by the NRC during Inspections 87-02 and 87-05.
The event report accurately described the circumstances of the occurrence and the lic-ensee's followup actions. The event report was submitted beyond the five-day time period allowed by 10 CFR 73.71 with the concurrence from the resident inspector, due to the nature of the event and the licensee's original intent to obtain further information.
No inadequacies were identified.
The inspector reviewed the licensee's response actions for an event that occurred at 9:49 a.m. on February 19, 1987 that involved the security hardware and caused a moderate loss of security effectiveness.
The licensee took prompt compensatory measures. The matter was reported to the NRC Duty Officer as a one-hour reportable event due to the conservative criteria in the licen-see's security procedures. The event was reportable within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in ac-cordance with 10 CFR 73.71. The hardware problem was corrected by 2:14 p.m.
on February 19, 1987. Although no specific cause for the February 19th fail-ure was identified, the inspector noted that hardware upgrades scheduled for completion by the end of March, 1987 would address the root cause of this and previous similar events. The inspector reviewed the licensee's description of the event reported in security event report No. 87-02 dated February 24, 1987.
No inadequacies were identified.
5.0 Operational Status Reviews Plant tours were conducted routinely to review activities in progress and to verify compliance with regulatory and administrative requirements. Tours of accessible plant areas included the control room, reactor building, turbine building, diesel generator rooms, and the protected area.
Radiation controls were reviewed in areas toured to verify access control barriers, postings and radiological controls were appropriate.
Plant housekeeping conditions and shift staffing were reviewed.
Shift logs and records were reviewed to deter-mine the status of plant conditions and the changes in operational status.
Actions taken by plant personnel during periods when equipment was inoperable were reviewed to verify that: technical specification limits were met; alter-nate surveillance testing was completed satisfactorily; and, equipment return to service upon completion of repairs was proper. This review was completed for the following items: (1)
"B" diesel generator inoperable from 9:35 a.m.
to 1:20 p.m. on February 4, 1987 due to faulty relief valve on air compressor discharge; (2) "B" diesel generator removed from service for preventive main-tenance on February 17, 1987; "A"/"B" toxic gas monitor maintenance on Febru-ary 17, 1987; and (3) "A" diesel generator inspection from 2:45 p.m. to 4:10 p.m. on February 27, 1987 for potential fuel line leaks.
Items that received additional review are discussed further belo.
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5.1 Safety System Review The residual heat removal, core spray, residual heat removal service water, high pressure coolant injection, service water, reactor core isolation cooling, standby liquid control and standby gas treatment systems were reviewed to verify that the systems were properly aligned and fully operational in the standby mode.
The review included verifi-cation that (1) accessible major flow path valves were correctly posi-tioned; (2) power supplies were energized, (3) lubrication and component cooling was proper, and (4) components were operable based on a visual inspection of equipment for leakage and general conditions.
No inade-quacies were identified.
5.2 Main Steam Line Radiation Monitor Alarm Setpoint During the routine monthly functional test of the main steam line radi-ation monitors per OP 4511.17 on February 17, 1987, the licensee deter-mined that the alarm setpoint for the
"C" monitor was higher than allowed at 198 mR/hr.
Technical Specification 3.1 and 3.2 require that the alarm setpoint be maintained 1.5 times full power background of 130 mR/hr, or 195 mR/hr, to alert the operator of abnormal primary coolant radiation levels. The technical specifications also require the monitors provide trip and isolation functions at 3 times normal background, which were found to be within specification during the February 17, 1987 test.
The inspector noted that no further actions were taken or required by operations personnel to trip the channels, since the RPS and PCIS trip setpoints were proper at 3 times background, and since the alarm function from the "B" trip system was operable via the operable "B" and "0" in-strument channels.
The licensee initiated a setpoint change request per VYAPF 0022.01 on February 17, 1987 to lower the alarm setpoints on the "A" and "C" main steam line monitors from 198 mR/hr to 180 mR/hr, which was less than 1.5 times the background radiation level of 130 mR/hr. No inadequacies were identified. However, this item is unresolved pending inspector review of the action to complete the setpoint change, and subsequent review of system performance based on the calibration and functional test results for the channels (UNR 87-04-01).
5.3 Inadvertent Toxic Gas System Initiation The "A" channel of the toxic gas monitoring (TGM) system tripped inad-vertently at 8:30 a.m. on February 17, 1987 due to a malfunction that occurred when the system performed an automatic span check of the carbon dioxide channel. The control room ventilation system isolated properly in response to the channel actuation, and the control room e:nergency breathing air system operated as required.
Plant operators isolated the
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faulty channel after verifying that the redundant channel in the "B" train did not show abnormal carbon dioxide levels, and restored the ventilation system to service.
l The toxic gas system was restored to an operable status for the remaining operable trip functions.
The toxic gas system has operability require-ments incorporated in the technical specifications.
Technical Specifi-cation 3.2.J and Table 3.2.7, allow plant operation with inoperable channels for each of the five toxic gases monitored by the system.
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inoperable equipment cannot be restored within 30 days following the loss of one channel, or 7 days after the loss of both channels, a special report must be submitted to the NRC within 14 days.
The operations shift supervisor initiated maintenance request (MR) 86-0301 to repair the "A" monitor and allowed I&C personnel to start work on the channel prior to completing the MR processing.
Following NRC in-spection of the maintenance activity (see Section 11 below) which re-vealed that work had been done to replace the flow meters on the."B" TGM channel as well as the "A", the shift supervisor initiated a maintenance request for the "B" channel, and declared the "B" channel operable at 2:35 p.m. after the successful completion of a system calibration. The
"A" TGM was declared operable at 3:10 p.m. on February 18, 1987 following a 24-hour observation period to monitor a newly installed carbon dioxide detector for stability.
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Both TGM trains were declared inoperable at 1:35 p.m. on February 20, 1987 when shift personnel noted erratic operation on both carbon dioxide channels. The plant entered a seven-day LCO per Technical Specification 3.2.7 to either fix the system or submit a special report. Both TGM trains were left on-line with operable trip functions for the remaining
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l four toxic gases. The trains were returned to a fully operable status l
at 3:25 p.m. on February 24, 1987 following actions under MR 87-335 to i
adjust the purge header pressure from the nitrogen supply from 5 to 15 l
psig, which stabilized the purge flow to the carbon dioxide reference cells.
Except as discussed in paragraph 11.1 below in the maintenance area, no inadequacies were identified.
5.4 Stack Gas Monitor Alarm Setpoints During a review of plant status logs on February 11, 1987, the inspector noted that PRO 87-06 was written on February 11, 1987 regarding alarm setpoints for the stack gas monitors that were appropriate to meet the requirements of Technical Specification 3.9.2, but were nonconservative relative to the limits of 10 CFR 50.72 and 50.73. The Part 50 reporting criteria require NRC notification of stack (environmental) release rates in excess of two times the applicable maximum permissible concentrations (MPC) of the limits established in Appendix B to 10 CFR 20, when averaged over a one hour period.
Setpoint change request 87-01 was completed on
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F February 13, 1987 to make the following changes to the Stack Gas I and II alarm setpoints: the high setpoint changed from 1.0X10E4 counts per minute (cpm) to 8.0X10E3 cpm; and, the high-high setpoint changed from 1.0X10+5 cpm to 8.0X10E4 cpm.
The inspector interviewed licensee personnel, reviewed logs and records, and completed independent calculations for the monitor setpoints to ver-ify they were determined in accordance with the Offsite Dose Calculation Manual (00CM), Section 5.0, Revision No. 3 dated January 27, 1987. The inspector also reviewed the licensee's evaluation for PRO 87-06 dated February 20, 1987, which concluded that the event was not reportable to the NRC under 10 CFR 50.73. The inspector identified no inadequacies with the licensee calculations or conclusions regarding reportability, for the reasons discussed below.
Technical Specification (TS) 3.8.E.1.a requires that the dose rate due
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to radioactive material released from the site in gaseous effluents be limited so as not to exceed, for noble gases, 500 mrem /yr to the total body and 3000 mrem /yr to the skin. The stack gas monitors provide online measurement of the dose rate for noble gases, and are required by Tech-nical Specification Table 3.9.2 to be operable with an alarm / trip set-point established per the ODCM at a value that will assure the limits of Specification 3.8.E.1.a (and 3.8.K.1) will be met.
The ODCM provides the methodologies to calculate the monitor setpoints based on measurement of the isotopic mix from the offgas process stream, and establishes the setpoint such that the requirements of TS 3.8.A will be met. The ODCM further states that the setpoints should be established in the manner so calculated, or at some conservative valve below the limit to assure TS compliance, such as that which might be based on con-trolling release rates from the plant in order to maintain offsite air concentrations below 2X MPC when averaged over an hour.
The inspector noted that the setpoints based on the 2X MPC "reportability" limits are about 20% less than those derived to meet the TS 3.8.A limits.
The inspector noted further that the above approach will provide a con-servative setpoint to assure the TS 3.8.A release limits are not exceeded.
The licensee reviewed the stack releases since 1982 and determined that there have been no releases that approached the 2X MPC limits. The stack gas monitors would have alarmed at the TS 3.8.A values, but created the potential that a reportability requirement may have been missed for re-lease rates slightly above the 2X MPC limit. However, the 10 CFR 50.72 and 73 reportability limits need not be construed as operability limits for the purpose of satisfying the TS LCO.
No inadequacies were identified.
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6.0 Feedwater Leak Detection System Status The inspector reviewed the feedwater leakage detection system and the monthly performance summary provided by the licensee in accordance with letter FVY 82-105.
The licensee reported that, based on the leakage monitoring data re-ported as of January 31, 1987, there were no deviations in excess of 0.10 from the steady state value of normalized thermocouple readings, and no failures in the 16 thermocouples installed on the four feedwater nozzles.
No unac-ceptable conditions were identified.
By letter FVY 86-29 dated March 28, 1986, the licensee informed NRC:NRR of the determinations regarding the need to continue PT examinations of the nozzles per the schedule in NUREG 0619.
The licensee determined that, based on previous PT results, the ongoing ultrasonic examinations, and the tempera-ture monitoring program, further PT examinations are no longer necessary or justifiable for the expended man-rem exposures.
This matter is under review by NRC:NRR.
This item is unresolved pending completion of the NRC staff re-view of the licensee's request (UNR 87-04-02).
7.0 Contaminated Material in Sanitary Dumpster The inspector reviewed a potential reportable occurrence report dated February 26, 1987 regarding the discovery of radioactive material outside the radiation controlled area, but within the protected area.
Details of the incident and the licensee's response to the event were discussed with the acting Plant Health Physicist on February 27, 1987 and March 2, 1987.
During a routine survey of waste prior to release from the site on February 25, 1987, a licensee technician identified a bag of trash reading 0.6 mrem /hr in a dumpster located outside the turbine building.
The radioactivity came from about one-half pound of unused but contaminated powdex resin, with an isotopic content of Co-60 and Cs-137 in a 10-to-1 ratio (about 1E-2 and 1E-3 uCi/gm) respectively.
Since the isotopic contents were similar to the site dry activated waste (DAW) make-up and unlike the mix from any facility process stream, the licensee concluded the resin became contaminated after being spilled on a contaminated floor surface within the radiation controlled area.
Following discovery of this problem, licensee personnel returned the bag of trash to the radiation controlled area for disposal as radioactive waste.
Subsequently, the inspector toured plant areas where the trash was handled, and measured the activity from the resin at about 50,000 dpm using a licensee RM-14 survey instrument.
The trash bag containing the resin most likely came from an HP station in the turbine building loading bay where a sorting station is set up for HP person-nel to segregate " contaminated" trash from " clean" trash.
Under existing controls, once HP personnel frisk material for disposal into the "cleun" re-ceptacle set up at the sorting station, the bags are closed and disposed of in the dumpsters without further piece-wise checking for contamination.
The material purportedly became uncontrolled when it was improperly disposed of
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in the " clean" trash receptacle at the sorting station after being swept up from the condensate demineralizer pre-coat area on the 232 ft elevation of the turbine building.
The licensee's review of the event was in progress at the conclusion of the inspection period. Controls for trash disposal were tightened requiring a second frisk of " clean" waste sealed in plastic bags prior to release from the RCA and disposal in the dumpsters.
The augmented controls were issued in a memorandum to health physics technicians dated February 27, 1987. The licensee stated that the event would be reviewed with operations personnel, and that a further study would be completed to verify the suspected chemical action between clean, moist resin and a contaminated surface.
The inspector calculated, using the approximate specific activities listed above and a 230 gram quantity of material, that about 2 uCi of Co-60 and 0.2 uCi of Cs-137 were deposited outside the radiation controlled area. The presence of the material in the dumpster constituted no safety hazard to plant workers, and control of the material was re-established by the licensee prior to removal from the owner controlled area. Based on the above activities and quantities, the inspector identified no conditions contrary to the posting and labeling requirements of 10 CFR 20.
However, the presence of the material in the unrestricted area was contrary to the requirements of licensee proce-dure RP 0521, Area and Equipment Decontamination, in that the material ex-ceeded the release limits (1000 dpm/100 sq. cm for removable and 0.1 mrem /hr for fixed contamination) for the release of radioactive material from con-trolled areas.
Failure to meet the AP 0521 limits constituted a violation of Technical Specification 6.5.B.
The resident inspector reviewed the event and determined that the licensee's corrective actions appeared appropriate for the assumed scenario on how the material became uncontrolled, and further were adequate to prevent recurrence.
The licensee's corrective actions for previous events (see below) do not appear to apply to this event.
No violation will be issued, pending satis-factory completion of the corrective actions, since the licensee's response to this item meets all the requirements of 10 CFR 2, Appendix C.
The inspec-tor had no further comments on this item at the present time. However, the inspector noted that this event appeared to be an example of a recurrent prob-lem in the failure to control low level radioactive material - reference In-spection Items 84-04-01, 84-21-10, 84-24-03 and 85-25-03.
This item is unre-solved pending completion of the licensee's corrective actions referenced above, and subsequent review of this event and the licensee's program for material control during a subsequent radiological control program inspection (UNR 87-04-03).
8.0 Audit Program The inspector reviewed the status of QA audit findings for plant activities, and in particular, for the areas of health physics and operations. Audit reports for the operations area (86-01 dated September 19, 1986 and 85-01 dated March 27,1985) and health physics area (86-03 dated June 27, 1986 and
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85-01 dated November 19,1985) were reviewed, along with the summary status listing for all plant audit findings.
The inspector noted that the audits appeared technically adequate.
Based on discussions with operational quality group (0QG) members, the inspector noted further that plant personnel are generally risponsive to audit deficiencies and findings, except as r!cted below.
Recent 00G findings identified the failure of plant responses tr meet the manager of operations (M00) implementation directive for audits. About half (71) of the 1986 audit finding response actions were not completed in accord-ance with the M00 directive. Most of these were in the radiation protection area, which also had a large number of findings recently due to high quality, in-depth audits. Most deficient response items concern audit " observations",
which are auditor suggested changes or recommendations for improvement. The large number of overdue items is partly due to the failure of the present program to differentiate between deficiencies and observations, and assign priortized M00 implementation deadlines accordingly. Based on a memorandum from the M00 to the plant manager dated February 2,1987, the inspector noted that licensee management has taken the initiative to recognize this program deficiency and initiate corrective actions.
The inspector had no further comments on this area at the present time. The effectiveness of the licensee's audit program will be reviewed during subse-quent routine NRC inspections.
9.0 Staffing Changes The licensee announced the following staffing changes during the inspection period. Mr. R. Lodwick resigned from the position of operations support man-ager in the corporate office. Mr. E. Jackson retired from the company, vacat-ing the position of assistant to the president. Mr. D. Reid was selected for the position of operations support manager. Mr. R. Wanczyk was selected for the position of operations superintendent.
Mr. R. Pagodin was selected for the position of technical services superintendent.
Mr. Wanczyk and Mr. Pagodin completed an SR0 certification program in 1985. Mr. J. Sinclair was selected for the position of plant services supervisor.
Mr. Sinclair will fill the security supervisor function until a permanent replacement is assigned. No inadequacies were identified.
10.0 Followup of Worker Concerns The inspector received a telephone call on February 17, 1987 from a contractor employee working as a QC inspector at a facility in the Southeast. The worker provided information relating to his concerns regarding material issuance and drawing controls for assuring environmental qualification of electrical equipment at the subject facility.
The worker also expressed concerns regard-ing potential discriminatory actions against him by utility personnel for disagreeing with the recommended resolutions for the issues he raised.
Based on the worker's descriptions of the issues, the inspector noted that the activities could involve potential violations of regulatory requirements and that the matter should be pursued by the NRC. The individual requested
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that his identity be kept confidential.
The above information was referred to Region I supervision on February 17, 1987. After consulting with a Region I coecialist, the inspector contacted the individual by telephone on February 10 a to discuss the technical issues further, and to provide him with a con-tact name and phone number in NRC Region II. This matter was referred to NRC Region II for followup.
11.0 Maintenance Activities Maintenance activity was reviewed to determine the scope and nature of work
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done on safety related equipment and equipment referenced in the technical specifications.
The revibw confirmed:
the repair of safety related equipment received priority attention; technical specification limiting conditions for
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operation were met while components were out of service; performance of al-
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ternate _ safety related systems was not impaired; and, the activity was com-j pleted in accordance with established procedures and plant equipment controls.
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Maintenance activity associated with the following was reviewed to verify (where applicable) procedure compliance and equipment return to service, in-cluding operability testing.
MR 87-219, "A" Diesel Generator on 2/17/87 - lube oil makeup line
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MR 87-227, "A" Diesel Generator on 2/17/87 - jacket cooler flange leak
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l MR 87-228, "A" Diesel Generator on 2/17/87 - exhaust line gasket leak
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MR 87-301, "A" Toxic Gai Monitor on 2/17/87 - spurious initiation
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e MR 87-335, "A'/"B" Toxic Gas Monitor on 2/20/87 - nitrogen flow oscilla-l
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I No inadequacies were identified, except as noted below.
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11.1 MR 87-301, Toxic Gas Monitor (TGM)
The carbon dioxide channel for the "A" TGM was removed from service on February 17, 1987 after the channel inadvertently actuated and initiated protection for the control room ventilation system. The operations shift supervisor requested maintenance of the inoperable channel, initiated maintenance request 87-301, and allowed I&C personnel to commence work on the system prior to the completion of actions to " process the MR".
I&C personnel replaced flow tubes in the purge gas supply associated with
'the carbon dioxide (CO ) reference cell, as part of system modifications
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'is used as a "zero" reference for the CO content of air as part of the
automatic zero and span check performed by the TGM system. The flow meters were replaced with smaller capacity (0-10 cu. ft./hr) tubes to provide more stable control of the purge flow. The old flow meters had a capacity of 0-100 cu. ft./hr, which contributed to the monitor problems
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sincecthey were operated on the lower end of their range to provide the design purge flow of about 4 cu. ft./hr. Operation at the bottom of the range resulted in-erratic purge makeup operation and caused the output of the CO analyzer to gradually increase toward its trip setpoint during
the automatic zero/ span checks as impurities leaked into the reference cell.
The workers completed the replacement on the "A" monitors, placed it back on-line, and then took the
"B" TGM system off-line to replace a similar flow meter in the,C0 channel on that system. The workers also replaced
the carbon dioxide analyzer cell detector on the "A" TGM system, and connected a recorder to the channel output to monitor its performance for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to returning the channel to an operable status. The inspector determined the status of the maintenance work activity and the above scope of work tased on discussion with the workers at about 1:00 p.m. on February 17, 1987.
During a subsequent review of the system status with the ir.spector on i
February 17, 1987, the operations shift supervisor stated he was unaware
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' of the work on the "B" TGM system and had not authorized it.
The shift supervisor contacted the workers, reviewed the status of the equipment
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and re-established equipment controls in accordance with plant procedures at 2:35 p.m., including the complstion of functional testing to verify operability of the "B" train.
he' inspector determined that the shift supervisor and the workers had previously discussed the workers intention to install two new flow meters, but because of, inadequate communications, the shift superviso:' understood that the meters would be changed out on the "A" train. The workers believed they had concurrence from the shift supervisor to complete the action on both trains.
The toxic gas system is classified by the licensee as non-nuclear safety s
related, but has operability requirements. incorporated in the technical specifications. Technical Specification 3.2.J and Table 3.2.7 allow plant operation with inoperable channels for each of' the five toxic gases monitored by the system.
If inoperable equipment cannot be restored within 30 days following the logs of one cnannel," or seven days after the loss of both channels, a special report must be submitted to the NRC within 14 days. Based on the above, and in consideration tn t there are no alternate LC0 actions required by the technica1rspecificat' ions for periods when both TGM trains are inoperable, the inspector concluded that the failure to control TGM system work activity on February 17, 1987 had minimal safety significance.
Plant procedure AP 0025, Plant Equipment Control, Revision 5, dated March 31,_1986, establishes the administrative requirements that must be fol-lowed for the removal of safety related or tech dca' specification re-ferenced plant systems from service for maintenance or modifications.
AP 0025 requires tnat the operations shift supervisor authorize work activity, at least orally, prior to the removal of plant equipment from service.
Such authorization is required in order to assure, f.n general,
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that plant safety or performance capability is not compromised, and that the equipment is controlled per established work control procedures dur-
' s ing and upon completion of work activities, including the completion of post repair operability testing and verification of the proper return
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to an operable status.
The failure to meet the requirements of AP 0025 during the work on the "B" TGM system on February 17, 1987, although an isolated case of improper administrative control, is a violation of the requirements of Technical Specification 6.5.A (VIO 87-04-04).
The inspector reviewed maintenance activity since the start of 1987 and noted nine occasions during which the operations shift supervisor initi-
'ated work activity to repair faulty plant equipment, and allowed initi-g'
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dtion of the work efforts prior to " processing the MR".
The inspector further noted that AP 0021, Maintenance Requests, Revision 14, dated N
' April 7,1986 inclusive of DI 87-05, does not appear to allow "short-cutting",the work scope review and approval process in the manner ob-served on those occasions. The inspector noted that, as part of his re-sponsibilities for safe plant operation, the shift supervisor should be able to expedite the work control process as necessary when plant condi-tions would warrant such actions. However, the necessity for such ex-pediencies was not immediataly apparent in all the examples noted by the inspector. This item is unresolved pending further inspector review of the maintenance activity, and further review of the licensee's actions for conformance with the established controls (UNR 87-04-05).
12.0 Potential Problem Involving Moveable Incore Flux Mapping Systems
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The inspector reviewed the incore flux mapping system at Vermont Yankee for similarities to the designs used at other facilities where seismic failure concerns have been identified (reference Information Notice (IN) 85-45 and NRC Temporary Instruction 2500-16). The issue involves the potential inter-action between the seismic and non-seismic portions of the incore flux mapping system at facilities using an incore detector seal table with vertically-mounted flux mapping equipment located above the pressure boundary piping.
The inspector reviewed the licensee's response to IN 85-45 as documented in a VYAPF 0028.02 memorandum dated July 1, 1985.
The licensee concluded that the concerns expressed in the notice had no impact on plant operations or safety since the installation at Vermont Yankee was different than that de-scribed in IN 85-45.
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The inspector noted that there is no seal table used in the traversing incore probe (TIP) system at Vermont Yankee, and that equipment and components in the incore detector system are mounted horizontally both inside and outside the drywell. The inspector noted further that the licensee had addressea the TIP room as part of the response to the block wall concerns identified in IE
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Bulletin 80-11. Thus, the potential failure of incore detector pressure boundary tubing due to interaction with the drive and shield components is
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virtually non-existent.
Based on the above, the inspector concluded that the concerns identified in IN 85-45 are not applicable to Vermont Yankee.
13.0 Spent Fuel Pool Expansion By letter FVY 86-34 dated April 25, 1986, the licensee submitted Proposed Change No. 133 to the NRC requesting a change to the Technical Specification 5.5.D to allow reracking the spent fuel storage pool to accommodate storage of up to 2870 spent fuel assemblies. This matter has been the subject of a technical review by the NRC:NRR staff through a series of meetings and in the exchange of correspondence requesting the licensee to address questions needed to complete the staff's review. By letter dated December 23, 1986, NRR issued
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a Notice of Consideration of Amendment to Facility Operating License DPR-28, a Notice of Proposed No Significant Hazards Determination, and a Notice of Opportunity For Hearing on the matter.
Pending satisfactory completion of the associated technical and administrative reviews, NRC staff approval of the licensee's proposed plans would be issued as an amendment change.
The facility is currently licensed to store 2000 spent fuel assemblies in the pool using Programmed and Remote Inc. (PAR) modular storage racks with al-ternate cells lined with boron. The licensee's proposal is to replace the
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PAR racks with higher density Nuclear Energy Services (NES) racks, that would change the center to center clearances between cells from 7.0 inches to 6.218 inches. The pool presently holds 1322 spent fuel assemblies and, with the present storage capacity of 1690 afforded by the existing PAR racks in the pool, the pool can accomodate an additional 368 fuel assemblies, or exactly one full core diccharge. The licensee will discharge 136 fuel assemblies into the storage pool during the shutdown for the refueling and maintenance outage scheduled to begin in August, 1987. Based on the above capacities, the plant will lose full core offload capability at the plant startup in October, 1987 unless additional storage capacity is added to the pool.
In addition to re-racking the pool, the licensee plans to modify the discharge piping to the pool from the spent fuel pool cooling system by cutting off the discharge nozzles located at the bottom of the pool, resulting in cooling water dis-charge at two locations above the top of the spent fuel on the East and West ends of the pool.
The licensee believes that the activities to modify the cooling water dis-l charge piping and to insert one rack of the new design without exceeding the current license limit of 2000 could be accomplished under 10 CFR 50.59. The licensee stated further that reviews have concluded that operation with full core discharge capability is a conservative operational practice that is not mandated by his existing license conditions. These matters were referred to NRC Region I management for review and actio.
The inspector observed and reviewed the results of activities in the spent fuel pool in progress on February 11, 1987 to " map" the dose rates adjacent to an existing full spent fuel storage rack with a 10 X 10 array. The purpose of the measurements was to establish an underwater dose profile versus dis-tance from filled storage locations to gather data as part of the ALARA plan-ning process for the reracking job.
Dose rate data around a filled PAR rack were as follows: 2 ft above the top middle of an array - 370 R/hr; 2 ft above the top edge of a full array - 200'R/hr; about 3.5 ft out from and at 1/3rd up the side of the array - 50 R/hr; and, about 5 ft from the side of the array - 1.7 R/hr. No inadequacies were observed in the survey technique or results.
The inspector had no further comments on this area at the present time.
Lic-ensee actions to rerack the spent fuel pool will be reviewed further during subsequent routine inspections.
14.0 Management Meetings Preliminary inspection findings were discussed with licensee management peri-odically during the inspection. A summary of findings for the report period was also discussed at the conclusion of the inspection and prior to report issuance.